The present invention pertains to the fields of BCC (body centered cubic) crystal structure metals and alloys and their use in water reactor fuel rod cladding. It is especially concerned with niobium and niobium base alloys, and ferritic stainless steels having properties which minimize the adverse effects of pellet-clad interaction (PCI) in light water reactor fuel elements.
The use of cladding tubes made entirely of a high zirconium alloy has been the practice in the water reactor industry. Examples of common alloys used are Zircaloy-2, Zircaloy-4 and zirconium-2.5 w/o niobium. These alloys were selected based on their nuclear properties, mechanical properties and high-temperature aqueouscorrosion resistance.
The history of the development of Zircaloy-2 and 4, and the abandonment of Zircaloy-1 and 3 is summarized in: Stanley Kass, "The Development of the Zircaloys," published in ASTM Special Technical Publication No. 368 (1964) pp. 3-27. This article is hereby incorporated by reference. Also of interest with respect to Zircaloy development are U.S. Pat. Nos. 2,772,964; 3,097,094; and 3,148,055.
Most commercial chemistry specifications for Zircaloy-2 and 4, and zirconium 2.5 w/o niobium, conform essentially with the requirements published in ASTM B350-80, (for alloy UNS No. R60802, R60804 and R60901, respectively) for example. In addition to these requirements, the oxygen content for the Zircaloy alloys is required to be between 900 to 1600 ppm but typically is about 1200.+-.200 ppm.
It has been a common practice to manufacture Zircaloy cladding tubes by a fabrication process involving: hot working an ingot to an intermediate size billet; beta solution treating the billet or log; machining a hollow; high temperature alpha extruding the hollow billet to a hollow cylindrical extrusion; and then reducing the extrusion to substantially final size cladding through a number of cold pilger reduction passes, having an alpha recrystallization anneal prior to each pass. The cold worked, substantially final size cladding is then final annealed. This final anneal may be a stress relief anneal, partial recrystallization anneal or full recrystallization anneal. The type of final anneal provided, is selected based on the designer's specification for the mechanical properties of the fuel cladding.
One problem that has occurred in the use of fuel rods utilizing the aforementioned cladding has been the observation of cracks emanating from the interior surface of the cladding which is placed under additional stress by contact with a fractured, thermally expanding oxide fuel pellet. These cracks sometimes propagate through the wall thickness of the cladding destroying the integrity of the fuel rod and thereby allowing coolant into the rod and radioactive fission products to contaminate primary coolant circulating through the reactor core. This cracking phenomena, is generally believed to be caused by the interaction of irradiation hardening, mechanical stress and fission products, producing an environment conducive to crack initiation and propagation in zirconium alloys.
Niobium foils have been proposed for use as an independent, or free standing, liner inside of a Zircaloy cladding tube (U.S. Pat. Nos. 3,925,151 and 3,969,186). Vapor deposition of a thin niobium coating on the inside surface of Zircaloy cladding tubes has also been proposed (U.S. Pat. No. 3,969,186). In addition, the use of niobium as a thin, buried layer, bonded to, and located between two concentric layers of Zircaloy, or an inner layer of zirconium and an outer layer of Zircaloy has also been proposed (U.S. Pat. No. 4,045,288).
Other efforts to inhibit PCI related crack propagation have involved zircaloy fuel cladding tubes having a zirconium layer autogeneously bonded to their inside surface (U.S. Pat. Nos. 4,372,817; 4,200,492 and 4,390,497).
The present inventor has now proposed the following alloy barrier fuel cladding design as an improvement over the aforementioned designs.
It is submitted that wrought niobium, and wrought zirconium-niobium base alloys, will be particularly effective as a barrier to the propagation of PCI related cracks when they are metallurgically bonded in a layer of at least 0.003 inches in thickness to the inside surface of water reactor fuel cladding tubes composed of conventional reactor grade zirconium base alloys. The niobium base materials in accordance with the present invention contain up to about 59 w/o zirconium, up to about 0.25 w/o tantalum, up to about 300 ppm of oxygen, with the balance being niobium with incidental impurities. Within this broad composition range, there are three classes of materials as shown in Table I.
TABLE I ______________________________________ A B C ______________________________________ O &lt;100 ppm &lt;200 &lt;300 Ta &lt;2500 ppm &lt;2000 &lt;1500 Zr &lt;200 ppm 0.02-13 w/o 13-59 w/o Nb remainder with remainder with remainder with incidental incidental incidental impurities impurities impurities ______________________________________
Class A materials are essentially pure, low oxygen niobium. Class B materials are low oxygen, zirconium-niobium base alloys having a stable fully beta niobium phase microstructure. Class C materials are low oxygen, zirconium-niobium base alloys having a metastable fully beta phase microstructure.
In an alternative embodiment of the present invention a layer of a ferritic stainless steel may be substituted for the niobium base layer previously described.
The foregoing and other embodiments and advantages will become more apparent upon review of the drawing in conjunction with the detailed description of the invention which follows.